RC-12-0136, Response to NRC Request for Additional Information Request for Weld Recategorization and Use of Alternative to ASME Code Inspection Interval for VCSNS Third Ten Year Inservice Inspection

Response to NRC Request for Additional Information Request for Weld Recategorization and Use of Alternative to ASME Code Inspection Interval for VCSNS Third Ten Year Inservice Inspection

ML12262A330
Person / Time
Site:	Summer
Issue date:	09/13/2012
From:	Gatlin T South Carolina Electric & Gas Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RC-12-0136, RR-1II-08, TAC ME8966
Download: ML12262A330 (27)

Text

Thomas D. Gatlin Vice President, Nuclear Operations 803.345.4342 SE&G September 13, 2012 A SCANA COMPANY RC-1 2-0136 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION (TAC NO. ME8966) REQUEST FOR WELD RECATEGORIZATION AND USE OF ALTERNATIVES TO ASME CODE INSPECTION INTERVAL FOR VCSNS THIRD TEN YEAR INSERVICE INSPECTION (RR-III-08)

Reference:

1. Letter from T. D. Gatlin, SCE&G, REQUEST FOR WELD RECATEGORIZATION FOR VCSNS STEAM GENERATOR NOZZLE TO SAFE END WELDS [PRELIMINARY SUBMITTAL - RR-III-08],

dated June 21, 2012, ADAMS Accession No. ML12177A381

2. Letter from T. D. Gatlin, SCE&G, REQUEST FOR WELD RECATEGORIZATION AND USE OF ALTERNATIVES TO ASME CODE INSPECTION INTERVAL FOR VCSNS THIRD TEN YEAR INSERVICE INSPECTION (RR-III-08), dated July 27, 2012, ADAMS Accession No. ML12212A287 This letter provides South Carolina Electric & Gas Company's (SCE&G) response to the Request for Additional Information (RAI) documented by Nuclear Regulatory Commission (NRC) email from Robert Martin dated August 31, 2011 at 7:22 am regarding the Request For Weld Recategorization And Use Of Alternatives To ASME Code Inspection Interval For VCSNS Third Ten Year Inservice Inspection (RR-III-08) pursuant to 1 OCFR50.55a(g)(6)(ii)(F)(2) and 1 OCFR50.55a(a)(3)(ii). The proposed request would allow an alternative categorization of the primary steam generator (SG) nozzle to safe end welds as required by inspection Item G of ASME Boiler & Pressure Vessel Code (BPVC) Code Case N-770-1 with conditions as specified in 1 OCFR50.55a(g)(6)(ii)(F). Additionally, the request would allow the station to defer the requirements of a baseline inspection from being performed in the first outage following January 20, 2012, by declaring hardship or unusual difficulty without a compensating increase in the level of quality and safety. Responses to the requested information are provided in the Enclosure.

Virgil C. Summer Station -Post Office Box 88 -Jenkinsville, SC

• 29065. F (803) 345-5209

/

Document Control Desk CR-12-00556 RC-12-0136 Page 2 of 2 Should you have questions, please contact Mr. Bruce L. Thompson at (803) 931-5042.

Very truly yours, Thomas D. Gatlin JG/TDG/bj

Enclosure:

Request For Additional Information Attachment I:

ALARA Projected Dose Study c:

K. B. Marsh S. A. Byrne J. B. Archie N. S. Cams J. H. Hamilton R. J. White W. M. Cherry V. M. McCree R. E. Martin NRC Resident Inspector K. M. Sutton NSRC RTS (CR-12-00556)

File (810.19-2)

PRSF (RC-12-0136)

Document Control Desk Enclosure CR-12-00556 RC-12-0136 Page 1 of 4 Enclosure VIRGIL C. SUMMER NUCLEAR STATION UNIT NO.1 REQUEST FOR ADDITIONAL INFORMATION (TAC NO. ME8966)

Request For Weld Recateqorization And Use Of Alternatives To ASME Code Inspection Interval For VCSNS Third Ten Year Inservice Inspection (RR-1II-08)

By "RC-12-0097, (VCSNS) Unit 1, Request for Weld Recategorization for VCSNS [[system" contains a listed "[" character as part of the property label and has therefore been classified as invalid. Nozzle to Safe End Welds (Preliminary Submittal - RR-III-08)|letter dated June 21, 2012]] (Agencywide Documents Access and Management System, Accession No. ML12177A381) and supplemented by letter dated July 27, 2012 (ADAMS Accession No. ML12212A287), South Carolina Electric and Gas Company submitted a request for weld recategorization and use of alternatives to ASME code inspection interval for VCSNS third ten year inservice inspection. The changes would allow an alternative categorization of the primary steam generator (SG) nozzle to safe end welds and provide relief from the baseline inspection requirements to be performed during the first outage following January 20, 2012.

VCSNS is requesting an alternative categorization to inspection items specific to the six replacement Steam Generator Nozzle (Hot and Cold Leg) to Safe End-Welds which are Alloy 82/182 with an Alloy 152 inlay that was originally applied during fabrication. The replacement steam generators were manufactured in 1993 and installed during VCSNS outage RF8 in the fall of 1994. Based on the manufacturing process of the replacement steam generators and guidance from Code Case N-770-1, the VCSNS Steam Generator Nozzles' (Hot and Cold Leg) Safe End Welds should be categorized as Inspection Item G, "Uncracked Butt Weld Mitigated With An Inlay."

VCSNS also proposes to delay the volumetric and inside surface inspection beyond the third 10-year inservice inspection (ISI) interval. The delay of the inspection would allow the development of inspection strategies and methodologies. Code Case N-770-1 requires a surface examination to be performed from the inside surface and a volumetric examination performed from either the inside or outside surface. Each steam generator nozzle has its own unique geometry due to the final field machining processes used to mate the nozzles to the pipes. The six unique geometries would require multiple mockups to demonstrate proficiency from the outside diameter.

VCSNS does not own a mockup of these nozzles and does not have a qualified technique in-house to perform outside diameter or inside diameter exams on this configuration. The delay would allow the development of an inside examination strategy. VCSNS has also determined that performing an ultrasonic and eddy current or penetrant test manually would subject personnel to a hazardous work environment and high radiation exposure. SCE&G proposes to conduct the baseline inspection with a qualified process during the following outage RF-21 scheduled for Spring 2014.

Document Control Desk Enclosure CR-12-00556 RC-12-0136 Page 2 of 4 The following questions were developed as a result of the U.S. Nuclear Regulatory Commission (NRC) Staffs review of the Request For Weld Recategorization And Use Of Alternatives To ASME Code Inspection Interval For VCSNS Third Ten Year Inservice Inspection (RR-II-08).

1. Based on your statement "Currently there are no qualified vendors performing automated encoded ultrasonic or eddy current testing from the inside diameter of a steam generator nozzle," what are the unqualified aspects of performing an ultrasonic examination from the ID at this time? How have you verified no qualification can be completed prior to the fall outage?

[VCSNS Response]

The unqualified aspects of this exam are:

1)

There are no VCSNS Electric Power Research Institute (EPRI)

Performance Demonstration Initiative (PDI) ultrasonic qualified technicians for performing automated encoded eddy current testing.

2)

There are no VCSNS qualified procedures for performing automated encoded eddy current testing.

3)

There is no VCSNS qualified ultrasonic process equipment (scopes, transducers, scanners, etc.) for performing automated encoded eddy current testing.

Therefore, the only possible options available at this time would be to use VCSNS personnel to perform an unqualified manual ultrasonic examination. The station does not have an eddy current process that could be used to detect the location of the inlay. Therefore, light acid etching of the inside diameter of each Alloy 152 inlay weld and a liquid penetrant surface exam would be required. The radiological impacts of this exam are discussed in the answer to RAI #2.

VCSNS contacted EPRI Nondestructive evaluation (NDE) department to identify qualified vendors and qualification requirements per the Performance Demonstration Initiative (PDI). At the time of the initial relief request submittal (Reference 1 and 2) there were no qualified processes or personnel. As of August 2012, there is only one vendor that has an EPRI Performance Demonstration Initiative (PDI) qualified process using the inside diameter dry contact ultrasonic examination of a Phased Array method with qualified procedures, equipment and personnel (1 individual). The resources and equipment required for this inspection are extremely limited due to the emergent nature of the new requirements and there are only enough resources and equipment for inspection of one plant. That inspection is scheduled to occur November 2012 and the vendor believes they will be available later that month (not accounting for delays) and are currently open for contract in 2014. VCSNS

Document Control Desk Enclosure CR-12-00556 RC-12-0136 Page 3 of 4 is scheduled to begin fuel movement on October 20, 2012 and conclude refuel movement on October 29, 2012. Therefore, the safest opportunity for the station to be configured at half pipe would be during the Defueled Work Window scheduled for October 22, 2012 through October 27, 2012. Taking the station to half pipe would add a minimum of seven days to the outage duration.

VCSNS has sought out the possibilities of having another automated ultrasonic testing system manufactured and the availability of qualified personnel to support this exam. However, based on prior commitments to industry peers, vendor support and a very heavy outage season, resources are limited. The EPRI PDI qualification process requires plant specific procedures, mock ups, laser profiling of the inside diameter and scan coverage calculations be established to have a code acceptable process. The one qualified vendor has identified the following list of items that should be considered to prepare for the examinations at VCSNS:

" Steam generator primary nozzle drawings review.

Tooling design engineer and NDE Level III will verify conditions of diameter, thickness, weld location, weld and surface conditions/geometry, steam dam ring configuration, and manway parameters note any differences and determine the effect on existing tooling or NDE procedure.

Validate NDE procedure and exam area conditions are consistent with the EPRI Performance Demonstration Initiative (PDI) qualified procedure.

Make any required adjustments or modifications to tooling (if required)

Determine if any additional fit and function type mockup is required (if tooling modifications are required they should be tested)

" Scan Plan Development Station general examination planning (i.e. determine equipment setup locations, submit procedures, personnel certifications, equipment certifications for review and approval, planning and scheduling meetings, etc.)

VCSNS is requesting approval for alternative inspection requirements in accordance with paragraph 10CFR50.55a(a)(3)(ii) of this section by declaring hardship or unusual difficulty without a compensating increase in the level of quality and safety. Implementation of the alternative inspection requires a relief pursuant to 1 OCFR50.55a(a)(3)(ii) for the baseline performance for six primary Steam Generator (nozzle to safe-end) welds during the next scheduled refueling outage. VCSNS is requesting to delay the inspection to provide additional time to plan, qualify and implement the examinations from the inside surface. Based on vendor input and scheduling conflicts the opportunities to qualify and complete these exams during RF-20 are not achievable.

Document Control Desk Enclosure CR-12-00556 RC-12-0136 Page 4 of 4

2. Based on your statement, "Performing an ultrasonic and eddy current or penetrant test manually would result in personnel performing a difficult test from the inside of the steam generator," what is the estimated radiological dose expected in performing this examination in the Fall 2012 outage rather than the automated examinations in the Spring 2014 outage?

[VCSNS Response]

The dose impacts from performing a manual volumetric examination and visual surface examination on the steam generator nozzle welds (hot leg and cold leg) internally during the Fall 2012 outage is estimated to be between 43 to 77 Rem.

There are no qualified techniques or outlined procedures available for the performance of an internal inspection of the steam generator nozzle welds at this time. The times used to calculate the estimated dose are based on historical similarities and input from a VCSNS qualified ultrasonic inspector. The nominal dose rates applied for this study are 1.000 Rem/hr for the hot leg and 2.000 Rem/hr for the cold leg. These rates are based on historical measurements located at the center of the bowl and at the bottom ofthe bowl.

The manual volumetric examination by ultrasonic testing from inside the steam generator would most likely take 75 minutes per weld and result in a dose of 3.750 Rem/generator or 11.250 Rem total. The visual surface examination by dye penetrant testing from inside the steam generator is projected at 120 minutes per weld and would have an exposure of 6 Rem/generator or 18 Rem total.

The robotic methodology would be the best alternative for conducting the volumetric examination and surface examination on the steam generator nozzle safe end welds (hot leg and cold leg) from the inside diameter. In comparison of the two activities (manual vs robotic) the difference is the examination activity itself and administrative coverage where manual would result in approximately 29.250 Rem of dose in comparison to the estimated 0.378 Rem for the robotic methodology. The remainder of the 43 to 77 Rem is attributed to steam generator manway removal, surface preparation, acid etching, radiological barrier controls and administrative operations at half pipe. Therefore, the robotic examination would provide a total of 28.872 Rem savings in personnel exposure for just the inspection activity alone. If the examination was conducted during the scheduled eddy current testing of the steam generator tubes that is expected to be completed during RF21 (Spring 2014), the administrative dose would be shared with the two activities minimizing exposure even further.

The details of the study and supporting surveys are attached.

Document Control Desk Enclosure CR-12-00556 RC-12-0136 Page 1 of 21 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS)

DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 Attachment I ALARA Projected Dose Study

VC Summer Nuclear Station A

Health Physics ALARA The dose impact from performing an internal, manual volumetric examination and visual surface examination on the steam generator nozzle welds (hot leg and cold leg) could be on the order of 43 to 77 Rem. To better manage and control worker doses, a well outlined process, qualified workers, and station plan needs to be developed that should include benchmarking of other facilities and mockup training. There are no qualified techniques or outlined procedures available for the performance of an internal inspection of the steam generator nozzle welds at this time. The closest guidance available for comparison purposes would be the following station Quality Services Procedures:

" QSP-501 "Solvent Removable Liquid Penetrant Examination,"

" QSP-505 "Visual Examination,"

" QSP-513 "Manual Ultrasonic Examination for Piping Systems," and

" QSP-518 "Acid Etching."

The steps (breakdown of the manual process) and man-hours used in the dose estimations that follow are a projection based on jobs with historical similarities and input from a qualified inspector.

Manual Inspection Method:

The code inspections of the steam generator primary side (hot and cold legs) would require the plant to drain down to half pipe. This activity has not been scheduled for RF20 and there is no current additional work requiring station alignment to half pipe.

The half pipe condition would add a minimum of seven days to the outage duration with an increased dose to the outage scope of approximately 0.700 Rem/day (or 4.9 Rem total).

The set up and removal of the necessary radiological barriers and controls (0.900 Rem) and radiation protection job coverage (0.825 Rem) would result in an additional 1.725 Rem.

The steam generator manway covers would have to be removed to access the inspection points inside the steam generator. Historically, workers receive 0.450 Rem for conducting the manway removal.

The dose rates utilized to evaluate the work inside the steam generator are an effective dose rate established from historical radiation and contamination surveys during eddy current testing. The nominal dose rates applied for this study are 1.000 Rem/hr for the hot leg and 2.000 Rem/hr for the cold leg. These dose rates are based on historical RC-12-0136 Enclosure Page 2 of 21

VC Summer Nuclear Station ta CW" Health Physics ALARA measurements located at the center of the bowl and at the bottom of the bowl. The surveys from 2005 (RF15) and 2009 (RF18) are attached.

The steam generator nozzle would need to be prepped and potentially buffed for volumetric examination. The projected time it would take to buff a 31-inch diameter pipe is estimated at approximately 40 minutes per weld. The estimated dose for the hot leg is 0.667 Rem and for the cold leg 1.333 Rem, therefore, the expected exposure is 2.000 Rem per generator or 6.000 Rem total for the weld preparation.

The time to conduct a manual volumetric examination from the inside the steam generator is estimated to take 75 minutes per weld. The estimated dose for the hot leg is 1.250 Rem and for the cold leg 2.500 Rem, therefore, the expected exposure is 3.750 Rem per generator or 11.250 Rem total for the volumetric examination.

The time needed for the visual surface examination by dye penetrant testing (PT) from the inside the steam generator estimated to take 120 minutes per weld. The estimated dose for the hot leg is 2.000 Rem and for the cold leg is 4.000 Rem. Therefore, the expected exposure is 6.000 Rem per generator or 18.000 Rem total for the visual surface examination.

There is a potential of not knowing where the edge of the weld is located. The common practice of acid etching will need to be used to locate the edges of the weld. The etching process on a 31-inch diameter pipe is estimated to take 120 minutes. The estimated dose for the hot leg is 2.000 Rem and for the cold leg 4.000 Rem, therefore, the expected exposure is 6.000 Rem per generator or 18.000 Rem total if weld etching is required.

The total collective dose estimate for the manual volumetric examination and visual surface examination is 37.425 Rem with an additional 4.900 Rem for establishing and maintaining half pipe conditions. The overall impact to the refueling outage combined would be 42.325 Rem, but if acid etching is required, the dose could be as high as 60.325 Rem.

There are many variables that could easily increase personnel exposure in this evolution. The amount of time needed for weld surface preparation may vary due to the level of surface imperfections should they exist. Additionally, the installation of foreign material exclusion (FME) covers or bladders should be considered and installed to protect the system and personnel. There could be a variation to the time based on the use of delta suits utilized for personnel protection. The physical configuration of the piping, the lack of lighting and first time evolution of performing the inspection could impact the stay time.

RC-12-0136 Enclosure Page 3 of 21

VC Summer Nuclear Station A* IOHealth Physics ALARA A radiation work permit estimate is considered to be successfully planned when it is within plus or minus 25 percent of the estimate. (This methodology is consistent with INPO 05-008, "Guidelines for Radiological Protection at Nuclear Power Stations,"

Revision 1, January 2011.) Therefore, the estimated total collective dose could result in 43 Rem or as high as 77 Rem.

The amount of resources that would be required for this method also poses a genuine concern. Given the maximum collective dose estimated and authorizing an administrative exposure limit extension to 1.500 Rem per worker, the total number of workers needed to support and accomplish a manual inspection could be as high as 45.

The total number of workers takes into account an administrative buffer of 90% (1.350 Rem per worker) to ensure no worker exceeds his/her limit. The availability of this many qualified workers on short notice would be a challenge.

Robotic/Remote Inspection Method:

The best solution is to have this inspection performed remotely via robotics.

Historically, robotics has saved time and dose. Additionally, the collective dose received would be minimal if performed in conjunction with a scheduled steam generator eddy current inspection. All of the set-up and man-way work would need to be performed to support the eddy current inspection. Some of the steps involved with the manual inspection would also be part of the robotic inspection. For instance:

" Establishing half pipe conditions for an estimated dose of 4.900 Rem.

" Removal of steam generator manway covers for an estimated dose of 0.450 Rem.

" Set up and removal of radiological barriers and controls (0.900 Rem)

" Radiation protection job coverage is reduced from 0.825 Rem to 0.175 Rem due to the limited activity required within the generator.

The robotic inspection would also require installing and removing the robot, which is estimated at the duration of 2.5 minutes per weld. At the expected dose rates, the estimated dose for the hot leg is 0.042 Rem and for the cold leg 0.084 Rem, therefore, the expected exposure is 0.126 Rem/generator or 0.378 Rem total for robotic examination.

Additional exposure savings could be obtained based on the weld preparation requirements. However, since these requirements are not fully understood a dose estimate of 6.000 Rem will be applied. VCSNS also postulated that weld etching will RC-12-0136 Enclosure Page 4 of 21

A MUM COAWANY VC Summer Nuclear Station Health Physics ALARA not be required for the automated process therefore the projected dose was reduced to zero.

Summary:

The robotic methodology would clearly be the best alternative for conducting the internal volumetric examination and surface examination on the steam generator nozzle welds (hot leg and cold leg). Comparison of the two examination methods would result in significant exposure savings of 29.522 Rem.

Manual examination method dose estimate of 42.325 Rem Robotic examination method dose estimate of 12.803 Rem If the examination was conducted during the scheduled eddy current testing of the steam generator tubes, that is expected to be completed during RF21 (Spring 2014),

the administrative dose would be shared and minimized even further.

Step Manual Examination Robotic Examination Half Pipe Dose (Administrative) 4.900 Rem 4.900 Rem Setup & Demobilize 0.900 Rem 0.900 Rem Full HP Coverage 0.825 Rem 0.175 Rem Remove Manways 0.450 Rem 0.450 Rem Weld Preparation 6.000 Rem 6.000 Rem Volumetric Examination 11.250 Rem 0.000 Rem Surface Examination 18.000 Rem 0.000 Rem Robotic Equipment Setup 0.000 Rem 0.378 Rem Total Estimated Dose 42.325 Rem 12.803 Rem Weld Etching (contingent) 18.000 Rem 0.000 Rem Gross Estimated Dose 60.325 Rem 12.803 Rem Personnel Required (1.35 Limit) 45 10 Planning Error (125%) Max Dose 76.250 Rem 16.000 Rem RC-12-0136 Enclosure Page 5 of 21

VC Summer Nuclear Station A

AAW Health Physics ALARA Attachments:

1. RF15 Steam Generator Survey Spring 2005 RF15 (Spring 2005) - "B" Steam Generator Primary Channel Head Survey RF15 (Spring 2005) - "C" Steam Generator Primary Channel Head Survey RF15 (Spring 2005) - "C" Steam Generator Primary Channel Head Survey RF15 (Spring 2005) - "A" Steam Generator Primary Channel Head Survey RF15 (Spring 2005) - "A" Steam Generator Primary Channel Head Survey

2. RF18 Steam Generator Survey Fall 2009 RF18 (Fall 2009) - "C" Steam Generator Primary Channel Head Survey RF18 (Fall 2009) - "C" Steam Generator Primary Channel Head Survey RF18 (Fall 2009) - "B" Steam Generator Primary Channel Head Survey RF18 (Fall 2009) - "B" Steam Generator Primary Channel Head Survey RF18 (Fall 2009) - "A" Steam Generator Primary Channel Head Survey RF18 (Fall 2009) - "A" Steam Generator Primary Channel Head Survey

3. RF15 Steam Generator Access Point Spring 2005 RF 5 (Spring 2005) - "C" Steam Generator Primary Manways RB-428' EL RF15 (Spring 2005) - "B" Steam Generator Primary Manways RB-428' EL RF15 (Spring 2005) - "A" Steam Generator Primary Manways RB-428' EL RF15 (Spring 2005) - "C" Steam Generator Primary Manways RB-428' EL RC-12-0136 Enclosure Page 6 of 21

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CENTER DIVIDER PLATE CENTER TUBE SREET

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?723 BOTTOM OF BOWL 12" INSIDE MANWAY PLANE OF MANWAY IV FROM MANWAY

/5' 3' FROM MAN WAY CONTACT DIAPHRAGM INSTALLED 12" DIAPHRAGM INSTALLED le CONTACT BACKSIDE OF DIAPHRAGM REMOVED 12" BACKSIDE OF DIAPHRAGM REMOVED x1

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REMOVED MA-CONTA1WMATION flil4/100_CM2.-

COUNTER:

Kt SER.10J INIST. TYPE!&62/0, 54.116620dc I R~815 12t 29N COUNTER:

14 SER.f 11 4 INST. TYPE: (Z,-SR

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COMMThENTS: -Shc)K~ 4.1 SURVEYED BY:4 5 12 19 26 33 ft-MP0, Stq,-rP'f -A ~(

TIME 2/521CDATE;

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______14 21 r_____

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5 ______

• DENOTES CONTACT READlINGS, ALL OTHERS IN mR/hr VfNLESS OTHERWISE NOTED (Beta/Gamma)

"I' Steam Generator Primary Channel Head Survey SURVEY *7____________

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7 T4 21 281 3_____

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• DENO0TES CONTAC READINGS, ALL OTHERS IN mRt/hr UNLESS OTHERWISE NOTED (Beta/Gaxmma)

(Page 50 of 115)

RF15 (Spring 2005)

• DENOTES CONTACT READINGS, ALL OTHERS IN RC-12-0136 Enclosure Page 18 of 21

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7 F.*:'.': *!'. : *:" ". ' I':. *;

(Page 5i of 115)

RF15 (Spring 2005)

SURVEY I)

SIJP/RWP#__05-X POWER 4

CONTAMINATION IVT DPH/100 CM2 :

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Rr4-14 SER,.#7 £q51 INST. TYPE: RO-zo SER.9 L(053 i

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19 1L6 123 1__

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SURVEYED BY -TAn--v 3

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_L L2 19 26 3_3_

TIMEb:/*Z, DATE: g./ '

• 4 EN1 28 R

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• DENOTES CONTACT READINGS, ALL OTHERS IN imR/hir UNLESS OTHERWISE NOTED (Beta/Gamma)

RC-12-0136 Enclosure Page 19 of 21 RF15 (Spring 2005)

(Page 52 of 115)

CONTAMINATION III DPM/IDOCE42 :.___

COUNTER: &M/C SER.i #j~

INST. TYPE:.Ž.O SER.U 0

dz/

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14-1 AM TIME: 1730 DATE:f 613 i

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ALL OTHERS IN MthRr UNLESS OTHERWISE NOTED (Beta/Gammna)

RC-12-0136 Enclosure Page 20 of 21 (Page 53 of 115)

RF15 (Spring 2005)

.1 1

RC-12-0136 Enclosure Page 21 of 21